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The role of parallel heat transport in the relation between upstream scrape-off layer widths and target heat flux width in H-mode plasmas of NSTX.

Description: The physics of parallel heat transport was tested in the Scrape-off Layer (SOL) plasma of the National Spherical Torus Experiment (NSTX) [M. Ono, et al., Nucl. Fusion 40, 557 (2000) and S. M. Kaye, et al., Nucl. Fusion 45, S168 (2005)] tokamak by comparing the upstream electron temperature (T{sub e}) and density (n{sub e}) profiles measured by the mid-plane reciprocating probe to the heat flux (q{sub {perpendicular}}) profile at the divertor plate measured by an infrared (IR) camera. It is found that electron conduction explains the near SOL width data reasonably well while the far SOL, which is in the sheath limited regime, requires an ion heat flux profile broader than the electron one to be consistent with the experimental data. The measured plasma parameters indicate that the SOL energy transport should be in the conduction-limited regime for R-R{sub sep} (radial distance from the separatrix location) < 2-3 cm. The SOL energy transport should transition to the sheath-limited regime for R-R{sub sep} > 2-3cm. The T{sub e}, n{sub e}, and q{sub {perpendicular}} profiles are better described by an offset exponential function instead of a simple exponential. The conventional relation between mid plane electron temperature decay length ({lambda}{sub Te}) and target heat flux decay length ({lambda}{sub q}) is {lambda}{sub Te} = 7/2{lambda}{sub q}, whereas the newly-derived relation, assuming offset exponential functional forms, implies {lambda}{sub Te} = (2-2.5){lambda}{sub q}. The measured values of {lambda}{sub Te}/{lambda}{sub q} differ from the new prediction by 25-30%. The measured {lambda}{sub q} values in the far SOL (R-R{sub sep} > 2-3cm) are 9-10cm, while the expected values are 2.7 < {lambda}{sub q} < 4.9 cm (for sheath-limited regime). We propose that the ion heat flux profile is substantially broader than the electron heat flux profile as an explanation for this discrepancy in the far SOL.
Date: January 5, 2009
Creator: Ahn, J W; Boedo, J A; Maingi, R & Soukhanovskii, V A
Partner: UNT Libraries Government Documents Department

MARFE Stability and Movement in an ELMy H-mode NSTX Discharge

Description: The results of a comparison of Multifaceted Asymmetric Radiation From the Edge (MARFE) theory with experiment in the National Spherical Torus Experiment (NSTX) are presented. A variety of MARFE behavior was observed using a fast-framing camera. A basic MARFE theory was applied to NSTX Multi-Pulse Thomson scattering (MPTS) and charge-exchange recombination spectroscopy (CHERS) data. MARFE theory showed some limited agreement with experiment, but uncertainty in the separatrix location constrained the analysis. A method based on shifting iso-Te flux surfaces was used to estimate the separatrix location. The movements of MARFEs in NSTX are interpreted to result from diamagnetic heat flux driven drifts relative to the background plasma velocity and imply slowing edge poloidal rotation and/or changing edge profiles before a large ELM.
Date: February 23, 2009
Creator: F. Kelly, R. Maingi, R. Maqueda, J. Menard, S. Paul
Partner: UNT Libraries Government Documents Department

Infrared Camera Diagnostic for Heat Flux Measurements on NSTX

Description: An infrared imaging system has been installed on NSTX (National Spherical Torus Experiment) at the Princeton Plasma Physics Laboratory to measure the surface temperatures on the lower divertor and center stack. The imaging system is based on an Indigo Alpha 160 x 128 microbolometer camera with 12 bits/pixel operating in the 7-13 {micro}m range with a 30 Hz frame rate and a dynamic temperature range of 0-700 degrees C. From these data and knowledge of graphite thermal properties, the heat flux is derived with a classic one-dimensional conduction model. Preliminary results of heat flux scaling are reported.
Date: March 25, 2003
Creator: Mastrovito, D.; Maingi, R.; Kugel, H. W. & Roquemore, A. L.
Partner: UNT Libraries Government Documents Department

Investigation of density limit processes in DIII-D

Description: A series of experiments has been conducted in DIII-D to investigate density-limiting processes. The authors have studied divertor detachment and MARFEs on closed field lines and find semi-quantitative agreement with theoretical calculations of onset conditions. They have shown that the critical density for MARFE onset at low edge temperature scales as I{sub p}/a{sup 2}, i.e. similar to Greenwald scaling. They have also shown that the scaling of the critical separatrix density with heating power at partial detachment onset agrees with Borass` model. Both of these processes yield high edge density limits for reactors such as ITER. By using divertor pumping and pellet fueling they have avoided these and other processes and accessed densities > 1.5{times} Greenwald limit scaling with H-mode confinement, demonstrating that the Greenwald limit is not a fundamental limit on the core density.
Date: February 1, 1999
Creator: Maingi, R.; Mahdavi, M.A. & Petrie, T.W.
Partner: UNT Libraries Government Documents Department

Investigation of physical processes limiting plasma density in H-mode on DIII-D

Description: A series of experiments was conducted on the DIII-D tokamak to investigate the physical processes which limit density in high confinement mode (H-mode) discharges. The typical H-mode to low confinement mode (L-mode) transition limit at high density near the empirical Greenwald density limit was avoided by divertor pumping, which reduced divertor neutral pressure and prevented formation of a high density, intense radiation zone (MARFE) near the X-point. It was determined that the density decay time after pellet injection was independent of density relative to the Greenwald limit and increased non-linearly with the plasma current. Magnetohydrodynamic (MHD) activity in pellet-fueled plasmas was observed at all power levels, and often caused unacceptable confinement degradation, except when the neutral beam injected (NBI) power was {le} 3 MW. Formation of MARFEs on closed field lines was avoided with low safety factor (q) operation but was observed at high q, qualitatively consistent with theory. By using pellet fueling and optimizing discharge parameters to avoid each of these limits, an operational space was accessed in which density {approximately} 1.5 {times} Greenwald limit was achieved for 600 ms, and good H-mode confinement was maintained for 300 ms of the density flattop. More significantly, the density was successfully increased to the limit where a central radiative collapse was observed, the most fundamental density limit in tokamaks.
Date: December 1996
Creator: Maingi, R.; Mahdavi, M. A. & Jernigan, T. C.
Partner: UNT Libraries Government Documents Department

Recent H-mode density limit experiments on DIII-D

Description: A vast body of tokamak data is in good agreement with the empirical density limit scalings proposed by Hugill and Greenwald. These scalings have common puzzling features of showing no dependence on either impurity concentration or heating power, since the density limit is frequently correlated with a rapid rise of the edge radiation. Despite the resiliency of these scalings, several machines under restrictive conditions have operated at densities well above the predictions of these scalings, albeit with pellet injection and at varying degrees of confinement degradation. Furthermore, data from several machines display a weak dependence on heating power. These results cast doubt on the universal validity of both of these scalings. Nevertheless the fact remains that access to densities above Hugill-Greenwald scaling is very difficult. A number of theories based on radiative power balance in the plasma boundary have explained some but not all features of tokamak density limit behavior, and as ITER design studies recently brought to focus, a satisfactory understanding of this phenomenon is lacking. Motivated by a need for better understanding of effects of density and fueling on tokamak plasmas in general, the authors have conducted a series of experiments designed to identify and isolate physical effects suspected to be directly or indirectly responsible for the density limit. The physical effects being considered include: divertor power balance, MARFE, poloidally symmetric radiative instabilities, MHD instabilities, and transport. In this paper they first present a brief summary of the experimental results up to the writing of this paper. The remainder of the paper is devoted to a comparison of this data at the onset of the MARFE instability with predictions of theory and the implications of the results on access to densities beyond the Hugill-Greenwald limit.
Date: June 1997
Creator: Mahdavi, M.A.; Maingi, R. & Hyatt, A.W.
Partner: UNT Libraries Government Documents Department

Characteristics of the scrape-off layer in DIII-D high-performance negative central magnetic shear discharges

Description: In this paper the authors present measurements of the global power and particle balance in the high-performance phase of negative central magnetic shear (NCS) discharges and compare with reference VH-mode discharges. The principal differences observed are that NCS has a much lower fraction of the total input power flowing into the boundary, less core radiation, and larger rate of stored energy increase as a fraction of total power. Scrape-off layer (SOL) temperature and divertor heat flux profiles, and radiation profiles at the midplane, are similar to VH-mode. Due to the good core particle confinement and efficient fueling by neutral beam injection (NBI), with little gas puffing, the gas load on the walls and the recycling are very low during the NCS discharges. This results in a rate of density rise relative to beam fueling at the L to H transition time which is 1/3 of the value for VH transitions, which is in turn 1/2 that for L-to-ELMing-H-mode transitions.
Date: October 1, 1996
Creator: Lasnier, C.J.; Maingi, R. & Leonard, A.W.
Partner: UNT Libraries Government Documents Department

Comparison of wall/divertor deuterium retention and plasma fueling requirements on the DIII-D, TdeV, and ASDEX-upgrade tokamaks

Description: The authors present a comparison of the wall deuterium retention and plasma fueling requirements of three diverted tokamaks, DIII-D, TdeV, and ASDEX-Upgrade, with different fractions of graphite coverage of stainless steel or Inconel outer walls and different heating modes. Data from particle balance experiments on each tokamak demonstrate well-defined differences in wall retention of deuterium gas, even though all three tokamaks have complete graphite coverage of divertor components and all three are routinely boronized. This paper compares the evolution of the change in wall loading and net fueling efficiency for gas during dedicated experiments without Helium Glow Discharge Cleaning on the DIII-D and TdeV tokamaks. On the DIII-D tokamak, it was demonstrated that the wall loading could be increased by > 1,250 Torr-1 (equivalent to 150 {times} plasma particle content) plasma inventories resulting in an increase in fueling efficiency from 0.08 to 0.25, whereas the wall loading on the TdeV tokamak could only be increased by < 35 Torr-{ell} (equivalent to 50{times} plasma particle content) plasma inventories at a maximum fueling efficiency {approximately} 1. Data from the ASDEX-Upgrade tokamak suggests qualitative behavior of wall retention and fueling efficiency similar to DIII-D.
Date: June 1, 1996
Creator: Maingi, R.; Terreault, B. & Haas, G.
Partner: UNT Libraries Government Documents Department

The role of neutrals in the H-L back transition of high density single-null and double-null gas-fueled discharges in DIII-D

Description: The role of neutrals in triggering the H-L back transition in high density ELMing H-mode plasmas is explored in double-null (DN) and single-null (SN) divertors. The authors propose that the neutral particle buildup below the X-point may play an important role in triggering the H-L transition at high density. Neutral pressure in the private flux region is, in fact, significant near the H-L backtransition. High density formation inside the separatrix near the X-point may also be a factor in triggering the H-L backtransition. They have observed that the ELMing H-mode density limit in SN divertors normally occurred at or near the H-L back transition. The radiated power coming from inside the separatrix at the H-L transition did not appear sufficient by itself to produce this back transition, since it is only {approximately}15--30% of P{sub in}. Poloidally-localized neutrals may explain two important differences in SN and double-null (DN) plasmas near their respective H-L backtransitions. First, electron pressure along the separatrix between the X-point and the outboard strike point decreased only modestly for DN divertors, even at densities comparable to the Greenwald density limit {bar n}{sub e,G}, in contrast to SN plasmas. Second, no divertor (or core) MARFEs were detected in the DNs as {bar n}{sub e} approaches {bar n}{sub e,G}, in contrast to SNs, where divertor MARFEs can form at {bar n}{sub e}/{bar n}{sub e,G} as low as {approximately}0.6. High X-point DNs achieved density limits well above those of comparably-prepared SNs, e.g., {bar n}{sub e}/n{sub e,G} {approx} 0.9--1.0 for DNs versus 0.75--0.80 for SNs. These differences result from a lower neutral pressure in the private flux regions of DNs than in comparable SNs at the same {bar n}{sub e}, since neutrals impact both pressure balance and MARFEing behavior.
Date: August 1, 1998
Creator: Petrie, T. W.; Maingi, R. & Porter, G. D.
Partner: UNT Libraries Government Documents Department

Density limit studies on DIII-D

Description: The authors have studied the processes limiting plasma density and successfully achieved discharges with density {approximately}50% above the empirical Greenwald density limit with H-mode confinement. This was accomplished by density profile control, enabled through pellet injection and divertor pumping. By examining carefully the criterion for MARFE formation, the authors have derived an edge density limit with scaling very similar to Greenwald scaling. Finally, they have looked in detail at the first and most common density limit process in DIII-D, total divertor detachment, and found that the local upstream separatrix density (n{sub e}{sup sep,det}) at detachment onset (partial detachment) increases with the scrape-off layer heating power, P{sub heat}, i.e., n{sub e}{sup sep,det} {approximately} P{sub heat}{sup 0.76}. This is in marked contrast to the line-average density at detachment which is insensitive to the heating power. The data are in reasonable agreement with the Borass model, which predicted that the upstream density at detachment would increase as P{sub heat}{sup 0.7}.
Date: August 1, 1998
Creator: Maingi, R.; Mahdavi, M.A. & Petrie, T.W.
Partner: UNT Libraries Government Documents Department

Divertor particle exhaust and wall inventory on DIII-D

Description: Many tokamaks achieve optimum plasma performance by achieving low recycling; various wall conditioning techniques including helium glow discharge cleaning (HeGDC) are routinely applied to help achieve low recycling. Many of these techniques allow strong, transient wall pumping, but they may not be effective for long-pulse tokamaks, such as the International Thermonuclear Experimental Reactor (ITER), the Tokamak Physics Experiment (TPX), Tore Supra Continu, and JT-60SU. Continuous particle exhaust using an in-situ pumping scheme may be effective for wall inventory control in such devices. Recent particle balance experiments on the Tore Supra and DIII-D tokamaks demonstrated that the wall particle inventory could be reduced during a given discharge by use of continuous particle exhaust. In this paper we report the first results of wall inventory control and good performance with the in-situ DIII-D cryopump, replacing the HeGDC normally applied between discharges.
Date: June 1, 1995
Creator: Maingi, R.; Wade, M.R. & Mioduszewski, P.K.
Partner: UNT Libraries Government Documents Department

NSTX Report on FES Joint Facilities Research Milestone 2010

Description: Annual Target: Conduct experiments on major fusion facilities to improve understanding of the heat transport in the tokamak scrape-off layer (SOL) plasma, strengthening the basis for projecting divertor conditions in ITER. The divertor heat flux profiles and plasma characteristics in the tokamak scrape-off layer will be measured in multiple devices to investigate the underlying thermal transport processes. The unique characteristics of C-Mod, DIII-D, and NSTX will enable collection of data over a broad range of SOL and divertor parameters (e.g., collisionality ν*, beta β, parallel heat flux q||, and divertor geometry). Coordinated experiments using common analysis methods will generate a data set that will be compared with theory and simulation.
Date: March 24, 2011
Creator: Maingi, R.; Ahn, J.-W.; Gray, T. K.; McLean, A. G. & Soukhanovskii, V. A.
Partner: UNT Libraries Government Documents Department

A Fast Visible Camera Divertor-Imaging Diagnostic on DIII-D

Description: In recent campaigns, the Photron Ultima SE fast framing camera has proven to be a powerful diagnostic when applied to imaging divertor phenomena on the National Spherical Torus Experiment (NSTX). Active areas of NSTX divertor research addressed with the fast camera include identification of types of EDGE Localized Modes (ELMs)[1], dust migration, impurity behavior and a number of phenomena related to turbulence. To compare such edge and divertor phenomena in low and high aspect ratio plasmas, a multi-institutional collaboration was developed for fast visible imaging on NSTX and DIII-D. More specifically, the collaboration was proposed to compare the NSTX small type V ELM regime [2] and the residual ELMs observed during Type I ELM suppression with external magnetic perturbations on DIII-D[3]. As part of the collaboration effort, the Photron camera was installed recently on DIII-D with a tangential view similar to the view implemented on NSTX, enabling a direct comparison between the two machines. The rapid implementation was facilitated by utilization of the existing optics that coupled the visible spectral output from the divertor vacuum ultraviolet UVTV system, which has a view similar to the view developed for the divertor tangential TV camera [4]. A remote controlled filter wheel was implemented, as was the radiation shield required for the DIII-D installation. The installation and initial operation of the camera are described in this paper, and the first images from the DIII-D divertor are presented.
Date: June 19, 2007
Creator: Roquemore, A; Maingi, R; Lasnier, C; Nishino, N; Evans, T; Fenstermacher, M et al.
Partner: UNT Libraries Government Documents Department

Magnetic Field Line Tracing Calculations for Conceptual PFC Design in the National Compact Stellarator Experiment

Description: The National Compact Stellarator Experiment (NCSX) is a three-field period compact stellarator presently in the construction phase at Princeton, NJ. The design parameters of the device are major radius R=1.4m, average minor radius &lt;a&gt; = 0.32m, 1.2 {le} toroidal field (B{sub t}) {le} 1.7 T, and auxiliary input power up to 12 MW with neutral beams and radio-frequency heating. The NCSX average aspect ratio &lt;R/a&gt; of 4.4 lies well below present stellarator experiments and designs, enabling the investigation of high {beta} physics in a compact stellarator geometry. Also the NCSX design choice for a quasi-axisymmetric configuration aims toward the achievement of tokamak-like transport. In this paper, we report on the magnetic field line tracing calculations used to evaluate conceptual plasma facing component (PFC) designs. In contrast to tokamaks, axisymmetric target plates are not required to intercept the majority of the heat flux in stellarators, owing to the nature of the 3-D magnetic field footprint. The divertor plate design investigated in this study covers approximately one half of the toroidal extent in each period. Typical Poincare plots in Figure 1 illustrate the plasma cross-section at several toroidal angles for a computed NCSX high-beta equilibrium. The plates used for these calculations are centered in each period about the elongated cross-section shown in Figure 1a, extending to +/- {pi}/6 in each direction. Two methods for tracing the edge field line topology were used in this study. The first entails use of the VMEC/MFBE-2001 packages, whereas the second entails use of the PIES code with a post-processor by Michael Drevlak; the same field line integration routine was used to evaluate the equilibria for this comparison. Both inputs were generated based on the {beta}=4%, =iota=0.5 equilibrium computed from the final NCSX coil set. We first compare these two methods for a specific plate geometry, and ...
Date: June 12, 2006
Creator: Maingi, R; Kaiser, T; Hill, D N; Lyon, J F; Monticello, D & Zarnstorff, M C
Partner: UNT Libraries Government Documents Department

Modeling of DIII-D noble gas puff and pump experiments

Description: Previous DIII-D experiments that induced a D{sup +} flow in the scrap-off layer (SOL) showed that this flow increased the divertor concentration of extrinsically injected impurities (neon, argon). These impurity fueling and exhaust (or puff and pump) experiments raise a number of modeling issues: the effect of edge-localized modes (ELMs) in regulating impurity core accumulation; the particle balance of the extrinsic impurities; the relation between divertor and plenum enrichment; and the effect of features unique to the present DIII-D Advanced Divertor configuration, specifically, the localized back-conductance of D{sub 2} and impurities from the baffle plenum in the outboard divertor region. To aid in understanding the relations between these processes, models have been improved: for core impurity transport to include ELM effects, and for divertor models to treat helium, neon, and argon transport with DIII-D--specific configuration effects. The models have been used to analyze a series of experiments in which neon and argon were first continuously injected (in the divertor private flux region) for 1.5 s, and then exhausted by the DIII-D cryopumping system. Deuterium was puffed at rates of 80 Torr L/s and 150 Torr L/s from the midplane and the divertor private region in these experiments. Results of the simulations are given.
Date: August 1, 1997
Creator: Hogan, J.T.; Wade, M.; Maingi, R.; Owen, L.; Schaffer, M. & West, P.
Partner: UNT Libraries Government Documents Department

Enhancement of Mode-converted Electron Bernstein Wave Emission during NSTX H-mode Plasmas

Description: A sudden, threefold increase in emission from fundamental electrostatic electron Bernstein waves (EBW) which mode convert and tunnel to the electromagnetic X-mode has been observed during H-mode [high-confinement mode] transitions on the National Spherical Torus Experiment (NSTX) spherical torus plasma. The mode-converted EBW emission viewed normal to the magnetic field on the plasma midplane increases when the density profile steepens in the vicinity of the mode-conversion layer, which is located in the plasma scrape off. The measured conversion efficiency during the H-mode is consistent with the calculated EBW to X-mode conversion efficiency derived using edge density data. Calculations indicate that there may also be a small residual contribution to the measured X-mode electromagnetic radiation from polarization-scrambled, O-mode emission, converted from EBWs.
Date: August 20, 2001
Creator: Taylor, G.; Efthimion, P.C.; Jones, B.; LeBlanc, B.P. & Maingi, R.
Partner: UNT Libraries Government Documents Department

Overview of impurity control and wall conditioning in NSTX

Description: The National Spherical Torus Experiment (NSTX) started plasma operations in February 1999, In the first extended period of experiments, NSTX achieved high current, inner wall limited, double null, and single null plasma discharges, initial Coaxial Helicity Injection, and High Harmonic Fast Wave results. As expected, discharge reproducibility and performance were strongly affected by wall condition. In this paper, the authors describe the internal geometry, and initial plasma discharge, impurity control, wall conditioning, erosion, and deposition results.
Date: May 23, 2000
Creator: Kugel, H.W.; Maingi, R.; Wampler, W.; Berry, R.E. & al, et
Partner: UNT Libraries Government Documents Department

Lithium Wall Conditioning And Surface Dust Detection On NSTX

Description: Lithium evaporation onto NSTX plasma facing components (PFC) has resulted in improved energy confinement, and reductions in the number and amplitude of edge-localized modes (ELMs) up to the point of complete ELM suppression. The associated PFC surface chemistry has been investigated with a novel plasma material interface probe connected to an in-vacuo surface analysis station. Analysis has demonstrated that binding of D atoms to the polycrystalline graphite material of the PFCs is fundamentally changed by lithium - in particular deuterium atoms become weakly bonded near lithium atoms themselves bound to either oxygen or the carbon from the underlying material. Surface dust inside NSTX has been detected in real-time using a highly sensitive electrostatic dust detector. In a separate experiment, electrostatic removal of dust via three concentric spiral-shaped electrodes covered by a dielectric and driven by a high voltage 3-phase waveform was evaluated for potential application to fusion reactors
Date: May 23, 2011
Creator: Skinner, C. H.; Bell, M. G.; Friesen, F. Q. L.; Heim, B.; Jaworski, M. A.; Kugel, H. et al.
Partner: UNT Libraries Government Documents Department

The Dependence of H-mode Energy Confinement and Transport on Collisionality in NSTX

Description: Understanding the dependence of confi nement on collisionality in tokamaks is important for the design of next-step devices, which will operate at collisionalities at least one order of magnitude lower than in present generation. A wide range of collisionality has been obtained in the National Spherical Torus Experiment (NSTX) by employing two different wall conditioning techniques, one with boronization and between-shot helium glow discharge conditioning (HeGDC+B), and one using lithium evaporation (Li EVAP). Previous studies of HeGDC+B plasmas indicated a strong and favorable dependence of normalized con nement on collisionality. Discharges with lithium conditioning discussed in the present study gen- erally achieved lower collisionality, extending the accessible range of collisionality by almost an order of unity. While the confinement dependences on dimensional, engineering variables of the HeGDC+B and Li EVAP datasets differed, collisionality was found to unify the trends, with the lower collisionality lithium conditioned discharges extending the trend of increasing normalized confi nement time with decreasing collisionality when other dimension less variables were held as fi xed as possible. This increase of confi nement with decreasing collisionality was driven by a large reduction in electron transport in the outer region of the plasma. This result is consistent with gyrokinetic calculations that show microtearing and Electron Temperature Gradient modes to be more stable for the lower collisionality discharges. Ion transport, near neoclassical at high collisionality, became more anomalous at lower collisionality, possibly due to the growth of hybrid TEM/KBM modes in the outer regions of the plasma
Date: November 28, 2012
Creator: S.M.. Kaye, S. Gerhardt, W. Guttenfelder, R. Maingi, R.E. Bell, A. Diallo, B.P. LeBlanc and M. Podesta
Partner: UNT Libraries Government Documents Department

The Dependence of H-mode Energy Confinement and Transport on Collisionality in NSTX

Description: Understanding the dependence of confi nement on collisionality in tokamaks is important for the design of next-step devices, which will operate at collisionalities at least one order of magnitude lower than in present generation. A wide range of collisionality has been obtained in the National Spherical Torus Experiment (NSTX) by employing two different wall conditioning techniques, one with boronization and between-shot helium glow discharge conditioning (HeGDC+B), and one using lithium evaporation (Li EVAP). Previous studies of HeGDC+B plasmas indicated a strong and favorable dependence of normalized con nement on collisionality. Discharges with lithium conditioning discussed in the present study gen- erally achieved lower collisionality, extending the accessible range of collisionality by almost an order of unity. While the confinement dependences on dimensional, engineering variables of the HeGDC+B and Li EVAP datasets differed, collisionality was found to unify the trends, with the lower collisionality lithium conditioned discharges extending the trend of increasing normalized confi nement time with decreasing collisionality when other dimension less variables were held as fi xed as possible. This increase of confi nement with decreasing collisionality was driven by a large reduction in electron transport in the outer region of the plasma. This result is consistent with gyrokinetic calculations that show microtearing and Electron Temperature Gradient modes to be more stable for the lower collisionality discharges. Ion transport, near neoclassical at high collisionality, became more anomalous at lower collisionality, possibly due to the growth of hybrid TEM/KBM modes in the outer regions of the plasma.
Date: November 27, 2012
Creator: S.M.. Kaye, S. Gerhardt, W. Guttenfelder, R. Maingi, R.E. Bell, A. Diallo, B.P. LeBlanc and M. Podesta
Partner: UNT Libraries Government Documents Department

Physics Design of the National High-Power Advanced Torus eXperiment

Description: Moving beyond ITER toward a demonstration power reactor (Demo) will require the integration of stable high fusion gain in steady-state, advanced methods for dissipating very high divertor heat-fluxes, and adherence to strict limits on in-vessel tritium retention. While ITER will clearly address the issue of high fusion gain, and new and planned long-pulse experiments (EAST, JT60-SA, KSTAR, SST-1) will collectively address stable steady-state high-performance operation, none of these devices will adequately address the integrated heat-flux, tritium retention, and plasma performance requirements needed for extrapolation to Demo. Expressing power exhaust requirements in terms of P{sub heat}/R, future ARIES reactors are projected to operate with 60-200MW/m, a Component Test Facility (CTF) or Fusion Development Facility (FDF) for nuclear component testing (NCT) with 40-50MW/m, and ITER 20-25MW/m. However, new and planned long-pulse experiments are currently projected to operate at values of P{sub heat}/R no more than 16MW/m. Furthermore, none of the existing or planned experiments are capable of operating with very high temperature first-wall (T{sub wall} = 600-1000C) which may be critical for understanding and ultimately minimizing tritium retention with a reactor-relevant metallic first-wall. The considerable gap between present and near-term experiments and the performance needed for NCT and Demo motivates the development of the concept for a new experiment--the National High-power advanced-Torus eXperiment (NHTX)--whose mission is to study the integration of a fusion-relevant plasma-material interface with stable steady-state high-performance plasma operation. Such a device would not have a high-fluence NCT mission, but would advance the science and technology necessary to accelerate the NCT mission at reduced risk in a separate nuclear facility. For the NHTX mission, flexibility to test multiple divertor configurations and first-wall components is critical, and flexibility in plasma exhaust configuration and boundary shape is important for understanding the plasma-wall interaction. Sufficient profile control must be available to generate high-performance ...
Date: July 2, 2007
Creator: Menard, J; Goldston, R; Fu, G; Gorelenkov, N; Kaye, S; Kramer, G et al.
Partner: UNT Libraries Government Documents Department